Log InRegister
Quick Links : The Mindat ManualThe Rock H. Currier Digital LibraryMindat Newsletter [Free Download]
Home PageAbout MindatThe Mindat ManualHistory of MindatCopyright StatusWho We AreContact UsAdvertise on Mindat
Donate to MindatCorporate SponsorshipSponsor a PageSponsored PagesMindat AdvertisersAdvertise on Mindat
Learning CenterWhat is a mineral?The most common minerals on earthInformation for EducatorsMindat ArticlesThe ElementsThe Rock H. Currier Digital LibraryGeologic Time
Minerals by PropertiesMinerals by ChemistryAdvanced Locality SearchRandom MineralRandom LocalitySearch by minIDLocalities Near MeSearch ArticlesSearch GlossaryMore Search Options
Search For:
Mineral Name:
Locality Name:
Keyword(s):
 
The Mindat ManualAdd a New PhotoRate PhotosLocality Edit ReportCoordinate Completion ReportAdd Glossary Item
Mining CompaniesStatisticsUsersMineral MuseumsClubs & OrganizationsMineral Shows & EventsThe Mindat DirectoryDevice SettingsThe Mineral Quiz
Photo SearchPhoto GalleriesSearch by ColorNew Photos TodayNew Photos YesterdayMembers' Photo GalleriesPast Photo of the Day GalleryPhotography

Wilson, D.; Iacovides, H.; Tatli, E.; Stansbury, C.; Ferroni, P. (2025) Thermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD. Nuclear Engineering and Design, 439. doi:10.1016/j.nucengdes.2025.114117

Advanced
   -   Only viewable:
Reference TypeJournal (article/letter/editorial)
TitleThermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD
JournalNuclear Engineering and Design
AuthorsWilson, D.Author
Iacovides, H.Author
Tatli, E.Author
Stansbury, C.Author
Ferroni, P.Author
Year2025 (August)Volume439
PublisherElsevier BV
DOIdoi:10.1016/j.nucengdes.2025.114117Search in ResearchGate
Generate Citation Formats
Mindat Ref. ID18381503Long-form Identifiermindat:1:5:18381503:7
GUID0
Full ReferenceWilson, D.; Iacovides, H.; Tatli, E.; Stansbury, C.; Ferroni, P. (2025) Thermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD. Nuclear Engineering and Design, 439. doi:10.1016/j.nucengdes.2025.114117
Plain TextWilson, D.; Iacovides, H.; Tatli, E.; Stansbury, C.; Ferroni, P. (2025) Thermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD. Nuclear Engineering and Design, 439. doi:10.1016/j.nucengdes.2025.114117
In(2025) Nuclear Engineering and Design Vol. 439. Elsevier BV

References Listed

These are the references the publisher has listed as being connected to the article. Please check the article itself for the full list of references which may differ. Not all references are currently linkable within the Digital Library.

ANSYS ICEM CFD, 2021.
Batta, A., Class, A.G. (2017) CFD analysis of pressure drop across grid spacers in rod bundles compared to correlations and heavy liquid metal experimental data. Nuclear Engineering and Design, 312. 121-127 doi:10.1016/j.nucengdes.2016.08.040
Not Yet Imported: - journal-article : 10.1080/00295450.2023.2181043

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Cevolani, S., 1995. ENEA themohydraulic data base for the advanced water cooled reactor analysis, in: 1st Research Co-Ordination Meeting of IAEA CRP on Thermohydraulic Relationships for Advanced Water Cooled Reactors, Vienna.
Cheng, Shih-Kuei, Todreas, Neil E. (1986) Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles — Bundle friction factors, subchannel friction factors and mixing parameters. Nuclear Engineering and Design, 92 (2). 227-251 doi:10.1016/0029-5493(86)90249-9
Cheng, X., Batta, A., Bandini, G., Roelofs, F., Van Tichelen, K., Gerschenfeld, A., Prasser, M., Papukchiev, A., Hampel, U., Ma, W.M. (2015) European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems. Nuclear Engineering and Design, 290. 2-12 doi:10.1016/j.nucengdes.2014.11.007
Not Yet Imported: - journal-article : 10.13182/NT88-A35553

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Di Piazza, Ivan, Angelucci, Morena, Marinari, Ranieri, Tarantino, Mariano, Forgione, Nicola (2016) Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility. Nuclear Engineering and Design, 300. 256-267 doi:10.1016/j.nucengdes.2016.02.008
Doolaard (2015) CFD benchmark for a heavy liquid metal fuel assembly
Eddinger, S.A., Czerniak, L., Ferroni, P., Liao, J., 2018. Lead (Pb) (No. RT-TR-18-16), Materials Property Manual. Westinghouse.
Not Yet Imported: - journal-article : 10.1016/j.anucene.2010.01.020

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Ferroni, P., Tatli, E., Stansbury, C., Liao, J., 2023. The Westinghouse Lead Fast Reactor: overview and update on development program, in: International Congress on Advances in Nuclear Power Plants. Presented at the International Congress on Advances in Nuclear Power Plants, Gyeongiu, Korea.
Ge, Zhihao, Liu, Jiaming, Zhao, Pinghui, Nie, Xingchen, Ye, Minyou (2017) Investigation on the applicability of turbulent-Prandtl-number models in bare rod bundles for heavy liquid metals. Nuclear Engineering and Design, 314. 198-206 doi:10.1016/j.nucengdes.2017.01.032
Not Yet Imported: Journal of Heat Transfer - journal-article : 10.1115/1.2911398

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Kennedy, G., Tichelen, K.V., Doolaard, H., 2015. Experimental investigation of the pressure loss characteristics of the full-scale MYRRHA fuel bundle in the COMPLOT LBE facility, in: 16th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). Presented at the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Chicago.
Not Yet Imported: 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) - proceedings-article : 10.13182/NURETH20-40056

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Not Yet Imported: - book-chapter : 10.1016/B978-0-444-82463-9.50015-0

If you would like this item imported into the Digital Library, please contact us quoting Book ID 9780444824639
Not Yet Imported: - journal-article : 10.1063/1.1432693

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Not Yet Imported: Journal of Heat Transfer - journal-article : 10.1115/1.3687091

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Marinari, R., Di Piazza, I., Tarantino, M., Forgione, N. (2019) Blockage fuel pin simulator experiments and simulation. Nuclear Engineering and Design, 353. 110215pp. doi:10.1016/j.nucengdes.2019.110215
Martelli, Daniele, Forgione, Nicola, Di Piazza, Ivan, Tarantino, Mariano (2015) HLM fuel pin bundle experiments in the CIRCE pool facility. Nuclear Engineering and Design, 292. 76-86 doi:10.1016/j.nucengdes.2015.06.004
Not Yet Imported: - journal-article : 10.1016/j.anucene.2017.01.031

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Not Yet Imported: - journal-article : 10.2514/3.12149

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Merzari, E., Fischer, P., Yuan, H., Van Tichelen, K., Keijers, S., De Ridder, J., Degroote, J., Vierendeels, J., Doolaard, H., Gopala, V.R., Roelofs, F. (2016) Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly. Nuclear Engineering and Design, 298. 218-228 doi:10.1016/j.nucengdes.2015.11.002
Mikityuk, Konstantin (2009) Heat transfer to liquid metal: Review of data and correlations for tube bundles. Nuclear Engineering and Design, 239 (4). 680-687 doi:10.1016/j.nucengdes.2008.12.014
Pacio, J., Daubner, M., Fellmoser, F., Litfin, K., Marocco, L., Stieglitz, R., Taufall, S., Wetzel, Th. (2014) Heavy-liquid metal heat transfer experiment in a 19-rod bundle with grid spacers. Nuclear Engineering and Design, 273. 33-46 doi:10.1016/j.nucengdes.2014.02.020
Pacio, J., Daubner, M., Fellmoser, F., Litfin, K., Wetzel, Th. (2016) Experimental study of heavy-liquid metal (LBE) flow and heat transfer along a hexagonal 19-rod bundle with wire spacers. Nuclear Engineering and Design, 301. 111-127 doi:10.1016/j.nucengdes.2016.03.003
Pacio (2015) Nuclear Engineering and Design, Thermal-Hydraulics of Innovative Nuclear Systems Heat transfer to liquid metals in a hexagonal rod bundle with grid spacers: Experimental and simulation results 290, 27
Not Yet Imported: - journal-article : 10.1016/0017-9310(73)90033-1

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Not Yet Imported: - journal-article : 10.1002/zamm.19510310704

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Roelofs, F. (2020) Liquid metal thermal hydraulics: State-of-the-art and future perspectives. Nuclear Engineering and Design, 362. 110590pp. doi:10.1016/j.nucengdes.2020.110590
Roelofs (2019)
Roelofs, F., Dovizio, D., Uitslag-Doolaard, H., De Santis, D., Mathur, A., Mikuz, B., Shams, A. (2019) Core thermal hydraulic CFD support for liquid metal reactors. Nuclear Engineering and Design, 355. 110322pp. doi:10.1016/j.nucengdes.2019.110322
Savatteri (1986) Proc.12th Liquid Metal Boiling Working Group Results and Comparison of Dry-Out in Grid and Wire Spaced Bundles at Single- and Two-phase Flow , 164
Not Yet Imported: - journal-article : 10.1016/0045-7930(94)00032-T

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Trupp, A.C., Azad, R.S. (1975) The structure of turbulent flow in triangular array rod bundles. Nuclear Engineering and Design, 32 (1). 47-84 doi:10.1016/0029-5493(75)90090-4
Ushakov (1977) Teplofiz. Vys. Temp. Heat transfer to liquid metals in regular lattices of fuel elements 15, 1027
Voj (1973)
Not Yet Imported: - journal-article : 10.1016/j.ijheatfluidflow.2022.109092

If you would like this item imported into the Digital Library, please contact us quoting Journal ID
Not Yet Imported: - journal-article : 10.1016/0017-9310(69)90012-X

If you would like this item imported into the Digital Library, please contact us quoting Journal ID

See Also

These are possibly similar items as determined by title/reference text matching only.

Raj, M.Naveen, Velusamy, K., Maity, Ram Kumar (2016) Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle. Nuclear Engineering and Design, 303. 88-108 doi:10.1016/j.nucengdes.2016.04.008
Stauff, Nicolas; Franceschini, Fausto; Ramey, Kyle; Richards, Scott; Liao, Jun; Jonson, Martin; Ferroni, Paolo (2025) Core design and performance of the Westinghouse lead fast reactor with UO2 and MOX configurations. Nuclear Engineering and Design, 442. doi:10.1016/j.nucengdes.2025.114266
Huang, Zheng; Jiao, Miaoxin (2025) Design and pre-test analyses of an integral thermal–hydraulic facility for a prismatic gas-cooled micro reactor. Nuclear Engineering and Design, 439. doi:10.1016/j.nucengdes.2025.114097
Padmakumar, G., Velusamy, K., Prasad, B.V.S.S.S., Lijukrishnan, P., Pandey, G.K., Selvaraj, P. (2018) Thermal-hydraulic effects of inserts in a fast reactor fuel bundle. Nuclear Engineering and Design, 332. 226-237 doi:10.1016/j.nucengdes.2018.03.037
Hung, T.C., Dhir, V.K., Chang, J.C., Wang, S.K. (2011) CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor. Nuclear Engineering and Design, 241 (1). 425-432 doi:10.1016/j.nucengdes.2010.10.015
Berthoux, Marc, Cadiou, Thierry (2010) The thermal hydraulics in a rod bundle representative of the start-up core of the ALLEGRO Gas cooled Fast Reactor—Experimental and numerical approaches. Nuclear Engineering and Design, 240 (10). 3372-3386 doi:10.1016/j.nucengdes.2010.07.009
Pérez-Valseca, Alejandría D.; Lopez-Solis, Roberto; Juárez-Martínez, Luis Carlos (2025) Neutronics and thermal hydraulics multiscale simulation of a lead-cooled fast reactor. Nuclear Engineering and Design, 441. doi:10.1016/j.nucengdes.2025.114189
Liu, C.C., Ferng, Y.M. (2010) Numerically simulating the thermal–hydraulic characteristics within the fuel rod bundle using CFD methodology. Nuclear Engineering and Design, 240 (10). 3078-3086 doi:10.1016/j.nucengdes.2010.05.021
Petroski, Robert, Hejzlar, Pavel, Todreas, Neil E. (2009) Thermal hydraulic design of a liquid salt-cooled flexible conversion ratio fast reactor. Nuclear Engineering and Design, 239 (12). 2612-2625 doi:10.1016/j.nucengdes.2009.07.012
Panigrahi, Prasant Kumar, Velusamy, K. (2021) Thermal-hydraulic analysis of FAIDUS design for severe accident mitigation strategy in a medium size sodium-cooled fast reactor. Nuclear Engineering and Design, 379. 111222pp. doi:10.1016/j.nucengdes.2021.111222
Orosz, Gergely Imre, Magyar, Boglárka, Szerbák, Dániel, Kacz, Dániel, Aszódi, Attila (2022) ALLEGRO Gas cooled Fast Reactor Rod Bundle investigations with CFD and PIV method. Nuclear Engineering and Design, 400. 112062 doi:10.1016/j.nucengdes.2022.112062
 
and/or  
Mindat.org is an outreach project of the Hudson Institute of Mineralogy, a 501(c)(3) not-for-profit organization.
Copyright © mindat.org and the Hudson Institute of Mineralogy 1993-2025, except where stated. Most political location boundaries are © OpenStreetMap contributors. Mindat.org relies on the contributions of thousands of members and supporters. Founded in 2000 by Jolyon Ralph.
To cite: Ralph, J., Von Bargen, D., Martynov, P., Zhang, J., Que, X., Prabhu, A., Morrison, S. M., Li, W., Chen, W., & Ma, X. (2025). Mindat.org: The open access mineralogy database to accelerate data-intensive geoscience research. American Mineralogist, 110(6), 833–844. doi:10.2138/am-2024-9486.
Privacy Policy - Terms & Conditions - Contact Us / DMCA issues - Report a bug/vulnerability Current server date and time: August 27, 2025 18:24:39
Go to top of page